• 한국물환경학회지
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• 제20권1호
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• pp.78-85
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• 2004

As an ozone contactor, we newly adopted HJLR (High-performance Jet Loop Reactor) for the decolorization of Reactive black 5 and the mineralization of oxalic acid, which has been applied exclusively in biological wastewater treatments and well-known for high oxygen transfer characteristics. The ozonation efficiency for organic removals and ozone utilization depending on the mass transfer rate were compared to those of Stirred bubble column reactor, which was controlled by varing energy input in the HJLR and Stirred bubble column reactor. The results were as follows; first, the decolorization rate of Reactive black 5 in the HJLR reactor was nearly proportional to the increasing $k_La$. When the $k_La$ was increased by 25 % from $13.0hr^{-1}$ to $16.4hr^{-1}$, 30 % of the k' (apparent reaction rate constant) was increased from 0.1966 to $0.2665min^{-1}$ (Stirred bubble column; from 0.1790 to $0.2564min^{-1}$). Ozone transfer was found to be a rate-determining step in decolorizing Reactive black 5, which was supported by that no residual ozone was detected in all of the experiments. Second, the mineralization of oxalic acid was not always proportional to the increasing $k_La$ in the RJLR reactor. The rate-determining step for this reaction was OH(OH radical) production with ozone transfer, because residual ozone was always detected during the ozonation of oxalic acid in contrast with Reactive black 5. This result indicates that the increase of $k_La$ in the HJLR reactor is beneficial only when there are in ozone transfer limited regions. In addition, regardless of $k_La$, the mineralization of oxalic acid was nearly accomplished within 60 minutes. It was interpreted as that the longer staying of residual ozone by whirling liquid in the HJLR reactor contributed to an high ozone utilization(83-94%), producing more OR radicals.

• Nuclear Engineering and Technology
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• 제39권2호
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• pp.133-142
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• 2007

The Electric Power Research Institute (EPRI) demonstrated a method for monitoring the performance of instrument channels in Topical Report (TR) 104965, 'On-Line Monitoring of Instrument Channel Performance.' This paper presents the results of several models originally developed by EPRI to monitor three nuclear plant sensor sets: Pressurizer Level, Reactor Protection System (RPS) Loop A, and Reactor Coolant System (RCS) Loop A Steam Generator (SG) Level. The sensor sets investigated include one redundant sensor model and two non-redundant sensor models. Each model employs an Auto-Associative Kernel Regression (AAKR) model architecture to predict correct sensor behavior. Performance of each of the developed models is evaluated using four metrics: accuracy, auto-sensitivity, cross-sensitivity, and newly developed Error Uncertainty Limit Monitoring (EULM) detectability. The uncertainty estimate for each model is also calculated through two methods: analytic formulas and Monte Carlo estimation. The uncertainty estimates are verified by calculating confidence interval coverages to assure that 95% of the measured data fall within the confidence intervals. The model performance evaluation identified the Pressurizer Level model as acceptable for on-line monitoring (OLM) implementation. The other two models, RPS Loop A and RCS Loop A SG Level, highlight two common problems that occur in model development and evaluation, namely faulty data and poor signal selection

• 한국압력기기공학회 논문집
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• 제8권1호
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• pp.33-38
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute has established a small-scale gas loop for the performance test on VHTR components and recently has manufactured a medium-scale PHE prototype made of Hastelloy-X. A performance test on the PHE prototype is scheduled in the gas loop. In this study, high-temperature structural analysis modeling, and macroscopic thermal and structural analysis of the medium-scale PHE prototype by imposing the established displacement boundary constraints in the previous research were carried out under the gas loop test condition. The results obtained in this study will be compared with performance test results.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.444-447
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• 2008

A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor (PWR) or a heavy water power reactor (CANDU). There is an in-pile section (IPS) and an out-pile section (OPS) in this test loop. When HANARO is normally operated, the fuel loaded in the IPS has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operation condition of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. The pump can not continuously suck a fluid and not pressurize the fluid during a cold function test. To verify the flow characteristics of the MCWS, a flow net work analysis has been conducted. When the higher elevation pipelines wholly filled with coolant, it was confirmed through the analysis results that the pump pressurized the coolant normally. And the analysis results described the system characteristics with operation temperature and pressure variation satisfactorily.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2006년도 제37회 하계학술대회 논문집 B
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• pp.875-876
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• 2006

Superconducting fault currents(SFCLs) are expected to improve not only reliability but also stability of power systems. The analysis on current limiting operations of the flux-lock type SFCL, which consists of a flux-lock reactor wound an iron core and a YBCO thin film, was compared the open-loop with the closed-loop iron core of the subtractive polarity winding. In the SFCL, operation characteristics could be controlled by adjusting the inductances and the winding directions of the coils, then magnetic field induced in the iron core. The current limiting characteristics under the same experimental conditions were generated regardless of the iron core conditions. We confirmed that capacity of the SFCL was increased effectively by the closed-loop iron core. However, the power burden of the system could be lowered by the open-loop iron core.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.

• Nuclear Engineering and Technology
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• 제35권2호
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• pp.91-107
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• 2003

Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1125-1139
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• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2708-2713
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• 2007

The aim of this paper is further studies to achieve deeper understanding in this field. First investigate the influence of operating conditions and design parameters on the hydrodynamics and the mass transfer properties of a loop reactor. This paper provides a literature review on the ejectors applications in the mixing system. A number of studies are grouped and discussed in several topics such as the background, theory of ejector, mixing characteristics, optimization of the system. Since the high efficiencies reactor using ejector widely used in gas-liquid system, especially in a number of chemical and biochemical processes. This is due to their high efficiency in gas dispersion resulting in high mass transfer rate and low power requirements. Thus ejector has been applied to the mixing system. An investigation on hydrodynamics and mass transfer characteristics of gas-liquid ejector has been carried out using three-dimensional CFD modeling.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.